1. Field of Invention
The present invention relates generally to the treatment of radioactive waste, and in particular, to the separation and regeneration/reclamation of specific radioactive isotopes from radioactive waste materials.
2. Description of the Related Art
Radioactive isotopes are commonly found in radioactive waste materials from numerous devices and processes associated with the use of nuclear energy, such as for example devices and processes associated with nuclear power plants. In nuclear power plants, radioactive isotopes leak into the primary and secondary water loops of light water nuclear reactors. This leakage is an unavoidable result of the decay of nuclear fuel as well as the nuclear activation (through exposure to radiation from the reactor core) of reactor components. The capability to isolate and manage specific radioactive isotopes is necessary for clean, safe, and secure radioactive waste management, which in turn is essential for the safe and cost-effective use of nuclear energy.
The concentration of particular radioactive isotopes in waste materials generally determines the waste classification of the waste material (for example, Class A, Class B, or Class C in the United States; or low level waste, intermediate level waste, or high level waste [LLW, ILW, or HLW] in other countries). The waste classification of waste material in turn delineates the storage and disposal requirements for that waste material. As a rule, waste material that receives a higher classification (such as Class B or Class C) faces stricter storage and disposal requirements, is more costly to manage, and can be legally stored in fewer locations. Therefore, it is desirable to limit the volume and amount of waste material that receives a higher classification by separating or removing from that waste material those specific radioactive isotopes that drive waste classification. In this regard, particularly desirable are systems, methods and processes for the separation of Cs-137, Sr-90, Ni-63, Tc-99, Am-241, Co-58, Co-60, and several isotopes of Uranium. It would also be advantageous for the isotope-separation technology to facilitate and work with technology for the processing of those specific radioactive isotopes for long-term storage or disposal, as for example through solidification or vitrification.
One method for removing radioactive isotopes from a liquid carrier, such as for example water, is to pass the liquid carrying the radioactive isotopes through an ion exchange media, consisting essentially of a substantially insoluble matrix, typically in the form of small beads or granules, which collectively define a structure of pores on the surface of which are sites with easily trapped and released non-radioactive ions. As the liquid carrying the radioactive isotopes passes through the ion exchange media, the radioactive isotopes are exchanged with the non-radioactive ions in the matrix, thereby trapping the radioactive isotopes within the ion exchange media and retaining the radioactive isotopes therein. The result is a liquid carrying fewer radioactive isotopes and a “depleted” ion exchange media having the removed radioactive isotopes trapped therein in place of the non-radioactive ions.
Because the above-discussed ion exchange process traps radioactive isotopes in the ion exchange media, the depleted media itself becomes radioactive. Accordingly, once the ion exchange media becomes substantially fully, or even partially, depleted, the depleted ion exchange media typically must also be disposed of as radioactive waste. And, as discussed above, the concentration of particular radioactive isotopes in the depleted ion exchange media generally determines the waste classification of the depleted media, which in turn delineates the storage and disposal requirements for the depleted media. Accordingly, it is desirable to limit the amount of depleted ion exchange media that receives a higher classification (i.e., that receives a classification greater than Class A).